Physics:Divertor
In magnetic confinement fusion, a divertor or diverted configuration is a magnetic field configuration of a tokamak or a stellarator which separates the confined plasma from the material surface of the device. The plasma particles which diffuse across the boundary of the confined region are diverted by the open, wall-intersecting magnetic field lines to wall structures which are called the divertor targets, usually remote from the confined plasma. The magnetic divertor extracts heat and ash produced by the fusion reaction, minimizes plasma contamination, and protects the surrounding walls from thermal and neutronic loads.
The term divertor usually describes the magnetic configuration itself or the region between the confined plasma and the target. Sometimes divertor target and divertor are used interchangeably. For example, the ITER divertor refers to the heavily engineered plasma-facing components designed to handle the intense plasma-wall interactions foreseen.
History
The divertor was initially introduced during the earliest studies of fusion power systems in the 1950s. It was realized early on that successful fusion would result in heavier ions being created and left in the fuel (the so-called "fusion ash"). These impurities were responsible for the loss of heat, and caused other effects that made it more difficult to keep the reaction going. The divertor was proposed as a solution to this problem. Operating on the same principle as a mass spectrometer, the plasma passes through the divertor region where heavier ions are flung out of the fuel mass by centrifugal force, colliding with some sort of absorber material, and depositing its energy as heat.[1] Initially considered to be a device required for operational reactors, few early designs included a divertor.
When early long-shot reactors started to appear in the 1970s, a serious practical problem emerged. No matter how tightly constrained, plasma continued to leak out of the main confinement area, striking the walls of the reactor core and causing problems. A major concern was sputtering in reactors with higher power and particle flux density,[2] which caused ions of the vacuum chamber's wall metal to flow into the fuel and to cool it.
During the 1980s it became common for reactors to include a feature known as the limiter, which is a small piece of material that projects a short distance into the outer edge of the main plasma confinement area. Ions from the fuel that are travelling outwards strike the limiter, thereby protecting the walls of the chamber from this damage. However, the problems with material being deposited into the fuel remained; the limiter simply changed where that material was coming from.
This led to the re-emergence of the divertor, as a device for protecting the reactor itself. In these designs, magnets pull the lower edge of the plasma to create a small region where the outer edge of the plasma, the "Scrape-Off Layer" (SOL), hits a limiter-like plate. The divertor improves on the limiter in several ways, mainly because modern reactors try to create plasmas with D-shaped cross-sections ("elongation" and "triangularity") so the lower edge of the D is a natural location for the divertor. In modern examples the plates are replaced by lithium metal, which better captures the ions and causes less cooling when it enters the plasma.[3]
In ITER and the latest configuration of Joint European Torus, the lowest region of the torus is configured as a divertor,[4] while Alcator C-Mod was built with divertor channels at both top and bottom.[5]
Tokamak divertors
A tokamak featuring a divertor is known as a divertor tokamak or divertor configuration tokamak. In this configuration, the particles escape through a magnetic "gap" (separatrix), which allows the energy absorbing part of the divertor to be placed outside the plasma. The divertor configuration also makes it easier to obtain a more stable H-mode of operation. The plasma facing material in the divertor faces significantly different stresses compared to the majority of the first wall.
Stellarator divertors
In stellarators, low-order magnetic islands can be used to form a divertor volume, the island divertor, for managing power and particle exhaust.[6] The island divertor has shown success in accessing and stabilizing detached scenarios and has demonstrated reliable heat flux and detachment control with hydrogen gas injection, and impurity seeding in the W7-X stellarator.[7][8] The magnetic island chain in the plasma edge can control plasma fueling.[9] Despite some challenges, the island divertor concept has demonstrated great potential for managing power and particle exhaust in fusion reactors, and further research could lead to more efficient and reliable operation in the future.[10]
The helical divertor, as employed in the Large Helical Device (LHD), utilizes large helical coils to create a diverting field. This design permits adjustment of the stochastic layer size, situated between the confined plasma volume and the field lines ending on the divertor plate. However, the compatibility of the Helical Divertor with stellarators optimized for neoclassical transport remains uncertain.[11]
The non-resonant divertor provides an alternative design for optimized stellarators with significant bootstrap currents. This approach leverages sharp "ridges" on the plasma boundary to channel flux. The bootstrap currents modify the shape, not the location, of these ridges, providing an effective channeling mechanism. This design, although promising, has not been experimentally tested yet.[12]
Given the complexity of the design of stellarator divertors, compared to their two-dimensional tokamak counterparts, a thorough understanding of their performance is crucial in stellarator optimization. The experiments with divertors in the W7-X and LHD have shown promising results and provide valuable insights for future improvements in shape and performance. Furthermore, the advent of non-resonant divertors offers an exciting path forward for quasi-symmetric stellarators and other configurations not optimized for minimizing plasma currents.[13]
See also
References
- ↑ "RF Absorbers material types". www.masttechnologies.com. http://www.masttechnologies.com/rf-absorbers/. Retrieved 30 August 2015.
- ↑ "Fusrev". Archived from the original on 2014-01-10. https://web.archive.org/web/20140110163654/http://freespace.virgin.net/colin.windsor/fusrev/fusrev.htm. Retrieved 2014-01-10.] T N Todd and C G Windsor, Progress in Magnetic Confinement Fusion Research, Contemporary Physics, 1998, volume 39, number 4, pages 255-282
- ↑ "Limiters and Divertors" , EFDA
- ↑ Stoafer, Chris (14 April 2011). "Tokamak Divertor System Concept and the Design for ITER". http://www.apam.columbia.edu/courses/apph4990y_ITER/Divertor%20Presentation%20-%20Stoafer.pdf.
- ↑ "MIT Plasma Science & Fusion Center: Research>alcator>information". Archived from the original on 2012-06-17. https://web.archive.org/web/20120617042944/http://www.psfc.mit.edu/research/alcator/intro/info.html. Retrieved 2012-09-11. retrieved September 11, 2012
- ↑ Feng, Y et al. (2006). "Physics of island divertors as highlighted by the example of W7-AS". Nucl. Fusion 46 (8): 807–819. doi:10.1088/0029-5515/46/8/006. Bibcode: 2006NucFu..46..807F. https://iopscience.iop.org/article/10.1088/0029-5515/46/8/006.
- ↑ Schmitz, O et al. (2021). "Stable heat and particle flux detachment with efficient particle exhaust in the island divertor of Wendelstein 7-X". Nucl. Fusion 61 (1): 016026. doi:10.1088/1741-4326/abb51e. Bibcode: 2021NucFu..61a6026S. https://www.osti.gov/biblio/1814444.
- ↑ Effenberg, F et al. (2019). "First demonstration of radiative power exhaust with impurity seeding in the island divertor at Wendelstein 7-X". Nucl. Fusion 59 (10): 106020. doi:10.1088/1741-4326/ab32c4. Bibcode: 2019NucFu..59j6020E. https://hal.archives-ouvertes.fr/hal-03740994/file/w7x-impurity-seeding-power-exhaust-florian-effenberg.pdf.
- ↑ Stephey, L et al. (2018). "Impact of magnetic islands in the plasma edge on particle fueling and exhaust in the HSX and W7-X stellarators". Physics of Plasmas 25 (6): 062501. doi:10.1063/1.5026324. Bibcode: 2018PhPl...25f2501S. https://pubs.aip.org/aip/pop/article/25/6/062501/320008/Impact-of-magnetic-islands-in-the-plasma-edge-on.
- ↑ Jakubowksi, M et al. (2021). "Overview of the results from divertor experiments with attached and detached plasmas at Wendelstein 7-X and their implications for steady-state operation". Nucl. Fusion 61 (10): 106003. doi:10.1088/1741-4326/ac1b68. Bibcode: 2021NucFu..61j6003J.
- ↑ Morisaki, T et al. (2013). "Initial experiments towards edge plasma control with a closed helical divertor in LHD". Nucl. Fusion 53 (6): 063014. doi:10.1088/0029-5515/53/6/063014. Bibcode: 2013NucFu..53f3014M. https://iopscience.iop.org/article/10.1088/0029-5515/53/6/063014.
- ↑ Boozer, A.H. (2015). "Stellarator design". Journal of Plasma Physics 81 (6): 515810606. doi:10.1017/S0022377815001373. Bibcode: 2015JPlPh..81f5106B. https://www.cambridge.org/core/journals/journal-of-plasma-physics/article/abs/stellarator-design/520178F88588D5631663E191EF6C6CE2.
- ↑ Bader, Aaron (December 6, 2018). "Progress in Divertor and Edge Transport Research for Stellarator Plasmas". https://www.jspf.or.jp/jspf_annual2018/JSPF35/pdf/S8-4.pdf.
Further reading
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